ML20076N081

Direct Testimony 1 Re Environ Qualification Equipment Classification & Evacuation Time Estimates

ML20076N081
Person / Time
Site:	Seabrook
Issue date:	07/15/1983
From:	Patricia Anderson, Devincentis J, Macdonald J, Maidrano R, Merlino R, Merrill D, George Thomas PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:
Shared Package
ML20076N057	List: ML20076N049 ML20076N081 ML20076N103
References
NUDOCS 8307210191
Download: ML20076N081 (43)

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Filcd:-July 15, 1983 APPLICANTS' DIRECT TESTIMONY No. 1 Members of the Panel:

David N. Merrill George S. Thomas-John DeVincentis David-A. Maidrand Peter L.' Anderson James A. MacDonald Robert J. Merlino i

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Table of Contents Page

Introduction:

Members of the Panel . . . . . . 1 Environmental Qualification: Time Duration . . . 7 Classification Scheme . . . . . . . . . . . . 13 Evacuation Times . . . . . . . . . . . . . . . 18 ,

Exhibits:

Exhibit 1 (SS-REP 5 5.0 & App. A)

Exhibit 2 (" Evacuation '

Clear Time Estimates for Areas Near Seabrook Station" updated 1981, revised July 1983 Exhibit 3 (Qualifications of David A. Maidrand)

Introduction:

Members of the Panel David N. Merrill is the Executive Vice President of Public Service Company of New Hampshire. He has been employed by PSNH since 1949; he was elected a vice president in 1965 and was elected Executive Vice President in 1973. Mr. Merrill has responsibility for Engineering, Production, Fuel Procurement and Supply, Energy Management and Research, and all aspects of Seabrook contruction and licensing. Mr. Merrill's qualifications appear in Appendix 13A of the FSAR.

Mr. George S. Thomas is Vice President - Nuclear Production of Public Service Company of New Hampshire.

Mr. Thomas began employment with Yankee Atomic Electric Company in 1969, and with PSNH in 1980. Mr. Thomas has been delegated the responsibilities of the Executive Vice President insofar as they include the operation of Seabrook Station, and Mr. Thomas therefore has overall responsiblity for the operation and associated support of Seabrook Station from the PSNH corporate office.

Mr. Thomas's qualifications appear in Appendix 13A of the FSAR.

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. t Mr. John DeVincentis is the Seabrook Station Project Manager for Yankee Atomic Electric Company.

Mr. DeVincentis joined Yankee in 1963 and has carried several responsibilities for several nuclear projects continuously since that time except for the four years from 1975 to 1979, during which time he was assigned to New England Power Company in connection with its proposed units in Charlestown, Rhode Island. Mr.

DeVincentis' qualifications are set forth in Appendix C to the FSAR.

Mr. David Maidrand is a Senior Project Engineer in the Seabrook Project Department of Yankee Atomic Electric Company. Mr. Maidrand commenced his employment with Yankee Atomic in 1974 following 9 years employment with the New England Electric System. Mr.

Maidrand's qualifications are set forth In Exhibit 3' hereto.

Mr. Peter L. Anderson is the Lead Seabrook Systems Engineer. Mr. Anderson became employed by Yankee Atomic Electric Company in 1981, following eleven years experience with Maine Yankee Atomic Power Company. Mr.

Anderson is a senior systems engineer for Seabrook and his qualifications appear in Appendix 13C of the FSAR.

Mr. James A. MacDonald is the Manager of the Yankee Radiation Protection Group. Mr. MacDonald began employment with Yankee in 1970 and has been Manager of the Radiation Protection Group since 1973. Mr.

MacDonald's qualifications appear in Appendix 13C of the FSAR.

Robert J. Merlino is the president and co-founder of HMM Associates of Waltham, Massachusetts. HMM Associates has been actively involved in radiological emergency planning since'the firm was founded, approximately five years ago. To date, HMM Associates has compiled evacuation time estimate studies for fourteen nuclear power plant sites in various parts of the country. Mr. Merlino has directed all emergency planning related work conducted by HMM Associates for PSNH, including the evacuation time estimate study. A copy of Mr. Merlino's resume follows on the next three pages:

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ROBERT J. MERLINO I

Education B.S. Civil Engineering, Tuf ts University,1963 Certificate in Reactor Engineering, Bettis M.B.A.

Reactor Engineering School, 1965 Business Administration, Babson College, 1970 Summary of Experience Mr. Merlino has broad experience in emergency planning, nuclear licensing ano project management. A nationally known authority on emergency planning for nuclear facilities, he has been involveo in emergency planning projects for over ten years, and has been active in AIF ano EEI emergency planning activities.

He has appearea as an expert witness on emergency planning, before NRC Atomic Safety and Licensing Boards. He has served as. project manager for licensing activities for a number of facilities.

Experience 1978 -

Present HMM Associates, principal and project manager.

He has led emergency planning activities on behalf of several nuclear utilities. This has,

( included evacuation studies, plan and procedures writing for stations, state and local plans, and reviews and audits of various types. He provides frequent consultation to nuclear management matters.

utility executives on regulatory and Most recently has provided technical support to Arizona Public Service Co. and Florida Power and Light in meeting emergency preparedness requirements.

His involvement has included ase,jstance with plant emergency plans, colporate plans, procedures, and state and local plans. He has provided a computer model for accident dose calculation, conducted training sessions-in use of the model, and assisted in scenario development and exercise-evaluation.

1977-1978 Environmental Research & Technology, Inc.

and (ERT).

1973-1976 Mr. Merlino held.a number of positions at ERT ~ as Project Manager, Manager of Air Quality Programs and Manager of Nuclear Services Division. He directed projects involving site selection, air quality and meteorological monitoring, radiological impact

(: assessments investigations.

and environmental' baseline He represented clients at meetings with regulatory agencies and at public hearings.

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i ROBERT J. MERLINO

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Among the projects he oirected were:

o Performance of environmental studies, preparation of environmental impact report and permitting for a pulp and paper mill expansion.

e o Evacuation analyses for twc proposed nuclear power stations.

o A probabilistic analysis of loss of coolant accident doses.

1 o Development of computer models for calcu-lating atmospheric dispersion.

o Installation and operation of meteorological and air quality monitoring networks and data acquisition systems around several foscil-fired and nuclear power stations.

1 1976-1977 Tera Corporation, Senior Project Manager.

Performed and managed engineering and environmental studies. This included the development of a sea-breeze fumigation model for

( calculating ground level concentrations from stack releases.

1968-1973 Yankee Atomic Electric Co. , Manager of Safety Analysis. Responsibilities included nuclear power plant site evaluations and participation in and direction of preparation of site-related portions of safety analysis and environmental reports for four' nuclear power stations. Topics included land use, meteorology, population distribution, evaluation of potential hazards from nearby industrial and military facilities and radiological safety. Prepared testimony and participated extensively in public hearings as an

' expert witness before state and federal regulatory bodies.

1967-1968 Pioneer Service and Engineering Co. , Nuclear Engineer. Performed safety and analyses for nuclear power stations and developed design requirements for engineered safety systems.

1963-1967 U.S. Navy, Division of Naval Reactors, Staff Engineer. While on active duty, directed government contractors in areas of nuclear k- propulsion plant mechanical systems _ design and testing.

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l ROBERT J. MERLINO I -

Professional Affiliations / Registrations Registered Professional Engineer (Nuclear), State of California American Nuclear Society I.

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EP 3/82

l Environmental Qualification: Time Duration (NECNP Contention I.B.2)

As admitted by the Board, this contention reads as follows:

"The Applicant has not satisfied the requirements of GDC 4 that all equipment important to safety be environmentally qualified because it has not specified the time duration over which the equipment is qualified."

This contention can be confusing, because it does not distinguish between two discrete concepts: (i) how long the equipment can be run, under normal circumstances, without losing its ability to withstand the harsh accident environment (should an accident occur), and (ii) how long the equipment can withstand the harsh environment after the accident has occurred.

For ease of reference, we shall refer to these concepts (which together constitute the " qualified life" of an item of equipment, as defined in IEEE 323-1974) as the

" pre-accident qualification duration" and the " post-accident qualification duration." From NECNP's interrogatories, it appears that the focus of this contention is the " post-accident qualification l

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duration." See interrogatories 14 and 29 of "NECNP's First Set of Interrogatories and Request for Documents to Applicants on Contentions I.A.2, I.B.1, I.B.2 and I.C" (filed 10/13/82).

As a matter of fact, the environmental qualification time duration standard for Seabrook Station electrical equipment is as follows: as to pre-accident qualification duration, the equipment in question is qualified either to the life of the plant or some shorter period, and if a shorter period is specified, then the equipment must be replaced or requalified before the period elapses. As to p'ost-accident qualification duration, all equipment is qualified to withstand accident environmental conditions for one year (the conditions being those set forth in " Service Environment Chart", Figure 3.11(B)-1, at FSAR $ 3.11), and any equipment that cannot be qualified for ona year is then reviewed on a case-by-case basis to determine whether, for the particular duration that equipment is required to remain operational in the case of an accident in order to perform its safety function, a shorter period is l

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sufficient. This standard fully meets (and in substantial part it exceeds) the requirement as to equipment environmental qualification time duration of General Design Criterion 4.

One other point of potential confusion should be cleared up at the outset. Our response to Interrogatory No. 1 of NECNP's second set of interrogatories explained that our definition of the terms "important to safety" and " safety related" were the same and are used interchangeably to identify structures, systems, or components that perform a safety function. We further stated that there is no equipment designated "important to safety" but'not safety-related. The following discussion, therefore, refers to all equipment that must be environmentally qualified to assure it will perform its safety function. We use the term " safety related" to refer to all such equipment.

Seabrook was in the preliminary design stage when Reg. Guide 1.89 was issued. Reg. Guide 1.89, which endorsed IEEE 323-1974, provided guidance for the first time on the requirement that electrical equipment be qualified to withstand an accident environment after having been exposed to pre-accident conditions for a qualified duration. It was decided that all safety-related electrical equipment not supplied by Westinghouse under the Nuclear Steam Supply System

("NSSS") contract would be qualified to perform its safety function in the harsh environment. Rather than determine specific accident scenarios for each application, all equipment was specified for a 40-yeat normal life followed by one year post-accident conditions. This common post-accident duration permits generic qualification of identical equipment irrespective of the requirements of the actual application. There were several reasons we chose this method. First and most importantly, we were not "backfitting" an existing design. By that, we mean that we were not trying to show that an existing design was acceptable. We were designing and purchasing systems to meet the requirements of IEEE-323-1974 and Reg. Guide 1.89. We felt that we could eliminate any potential for error that might exist if one tried to identify specific operating durations and accident r

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environments for each piece of equipment. By this approach, we achieved the flexibility of locating equipment and interchanging like equipment among systems without jeopardizing equipment qualification.

We also knew that in most cases, vendors either qualified equipment !or in-containment accident conditions or for mild environments and there was little financial benefit to attempting to identify i

conditions which fell between these extremes.

The result is that much of our safety-related equipment will be qualified to operate longer and in environments that are more severe than is required by GDC 4.

As we proceed with our detailed review of the equipment qualification data packages, we may find equipment that cannot be qualified to operate for one year in the extreme environment. Should this situation 1

arise, we will identify the time duration over which l the equipment is required to perform its safety function, with a margin. That particular equipment will then be qualified to that specific duration. This

will be done on a case-by-case basis and will be so identified in the equipment qualification data file.

To date, we have not identified any equipment that cannot be qualified for one year of post-accident environmental conditions.

The NSSS safety-relt.ted electrical equipment is qualified using specific qualification times based on the accident scenarios for each specific equipment application. These are set forth in Table 3.11(N)-3 in FSAR $ 3.11. Applicability to the Seabrook conditions will be verified by comparing the qualification profile to ensure that the test profiles envelope the Seabrook profile. This qualification meets or exceeds GDC 4, Reg. Guide 1.89, and IEEE 323-1974.

For these reasons, the Seabrook environmental qualification program meets or exceeds the requirements of GDC 4.

1 Classification Scheme (Contention NECNP III.1 & NH-20)

Introduction The Seabrook Station Radiological Emergency F.

(SS-REP) includes a system for emergency recognition and classification as the basis far the activation of the Applicant's emergency response organization and the notification to and activiation of the emergency response organizations of the federal, state, and local authorities. Emergency conditions are categorized by the SS-REP into one of the following four emergency classes:

1. Unusual Event

2. Alert

3. Site Area

4. General These four emergency classes are the same as those incorporated into federal, state, and local radiological emergency plans and which govern their response to an emergency notification by Seabrook Station personnel. This uniformity in the classification system used by the Applicants with that

r-used by the federal, state, and local authorities is in accordance with 10 CFR $ 50.47(b)(4).

Classification System Descripton The four emergency classes cover a graded scale of severity from a potential degradation of plant safety margins at the Unusual Event leiel to substantial core degradation or melting with potential for loss of containment integrity at the General Emergency level.

An explanation of all four emergency classes to show this graded scale of emergency severity is given in Section 5.0 of the SS-REP.

That section, as well as its companion Appendix A of the SS-REP, was transmitted to the parties by PSNH letter to USNRC, SBN-525 entitled Emergency Classification System, dated June 27, 1983, and a copy is annexed to this testimony as Exhibit 1.1 The 8It has also been included in Amendment 49 of the FSAR, as indicated in that letter. It should be noted that Table A.5 in Exhibit 1 contains certain typgraphical corrections from the document transmitted on June 27, 1983.

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revised Section 5.0 describes the manner by which 4

emergency conditions are categorized into the four emergency classes (Subsections 5.3.1 through 5.3.4).

It also describes the use of the symptomatic approach being incorporated into the Emergency Operating Procedures (EOPs) by the Applicants as an aid to emergency recognition and classification by the operator. This is described in the revised Subsection 5.2 and is set forth in dotail in the information specified in the revised Appendix A.

This symptomatic approach being utilized in the development of the Seabrook Station EOPs is a result of the more than three year effort of the Westinghouse Owners Group (which includes Seabrook Station representation) to improve the methods by which EOPs were based and written and operators oriented and trained to respond to emergency conditions.

Improvement was jointly recognized by the industry and the NRC in the analysis of the TMI accident response.

The Critical Safety Functions (CSFs) (those functions that must be successfully maintained to insure adequate safety margins), and the symptoms that represent 15 -

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l various degrees of challenge to them, were identified to concentrate the operators' attention on that information that is most important in an emergency situation.

As the materials in SS-REP Section 5.0 and Appendix A demonstrate, the Applicant's emergency classification system fully utilizes this symptomatic approach to CSF challenges. Subsection 5.2 describes the concept of color-coded status trees associated with the symptomatic EOPs and Appendix A shows in detail the relationship between status trees associated with the i

five CSFs and the categorization of the condition into one of the four emergency classes. These CSF status trees are used to 1) monitor station safety status, 2) alert operators to potential emergency conditions, and

3) direct operators to appropriate CSF restoration procedures. These CSF status trees are available on the Safety Parameter Display System of the plant computer and are also avialable as hard copy for backup.

Additionally, Appendix A also shows that events that are not represented, at least initially, as a CSF

threat are categorized into the emergency classification system. These types of events are occurrences that are mainly external to the systems and equipment associated with the maintenance of CSFs. In this way, the various challenges to CSFs and all the other appropriate actual or potential emergency conditi,ons are incorporated into the Applicant's emergency classification system.

Conclusion The information provided to the Shift Superintendent to recognize emergency conditions and categorize them in accordance with the emergency classification system as describer' above and presented in detail in the referenced submittal is a means to smoothly transfer between emergency condition recognition and the categorization and classification step. The incorporation of the symptom-based CSF status tree approach used for the EOPs into this classification system aids in this process.

Evacuation Times (NECNP Contentions III.12 and .13)

This Board's Order of June 30, 1983, disposed of the entirety of this contention except for two respects. The Board has reframed the remaining aspects of the contention thus: -

"NECNP III.12/III.13 Evacuation Time Estimates "The evacuation time estimates provided by i Applicants in Appendix C of the Radiological Emergency Plan are deficient in failing to include an estimate of:

"1. the times for evacuation during adverse weather conditions deveoping on a busy summer weekend; and "2. the times for simultaneous evacuation of beach areas lying NE to SSE of the Seabrook site."

This testimony is limited to those two issues.

Attached hereto as Exhibit 2 is a study entitled

" Evacuation Clear Time Estimates for Areas Near l Seabrook Station" (updated 1981, revised July 1983).

This is in essentially the same format as, and uses the i

same data and computer program as, Appendix C to the SS-REP contained in the FSAR, and it is, in fact, the study of which Appendix C is a condensed version. The differences between Exhibit 2 and Appendix C are-i

threefold: ,

one, Exhibit 2 contains a somewhat fuller description of the evacuation time estimate methodology; two, Exhibit 2 contains an evacuation time estimate for simultaneous evacuation of the entire EPZ (under the same summer scenarios initially studied);

and three, Exhibit 2 contains an evacuation time estimate for simultaneous evacuation of the entire EPZ under the peak weekend population-adverse weather scenario.2 The " entire EPZ" scenarios for the original summer cases produce evacuation clear times of-6 hours, 5 minutes for the summer weekend-fair weather scenario and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 10 minutes for the summer weekday-fair weather scenario.

2This last item is an estimate compiled in response to this Board's Order of June 30, 1983. The originai

" entire EPZ" case was run at the time the study was first performed, but.it was not reported in SS-REP Appendix C because, at the time Appendix C was published, NUREG-0654/ FEMA-REP-1, Rev. O, did not require such a case and the effort was to portray that called for by FEMA-REP-1.

Exhibit 2 also contains a summer weekday-fair weather scenario for each. sector. These, too, were not included in Appendix C because they were not called for by FEMA-REP-1.

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The " entire EPZ" case for the summer weekend-l adverse weather scenario yields an evacuation clear time of 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />, 15 minutes.

All of the " entire-EPZ" cases account for an evacuation of the beach area from NE to SSE, plus all other areas of the EPZ, at the same time.

While we have included this last estimate (i.e.,

summer weekend-adverse weather), we wish to point out that, in our judgment, it overstates the time that a real life evacuation would take under peak population, adverse weatner conditions. The reason for this is that, in order to model this scenario, it was assumed that none of the people at the beach areas began to depart -- notwithstanding the degrading weather --

until the signal to evacuate was given. In real life, one of two things would happen: either (1) the weather would begin to degrade before the notification was given, in which case some people would begin to leave before the signal was given, or (2) the weather would not begin to degrade until after the evacuation notification was given, in which case the effects of adverse weather upon evacuation would not appear until l

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the evacuation was underway and at least partially completed. Either of these real life situations would produce lower evacuation times than the case actually modelled.

In all other respects, these estimates are based upon the same methodology, the same assumptions, and the same data as did those the results of which are described in SS-REP Appendix C.

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• SEAMOOK STATION i 4 "*

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bec: J.P. Cady J.E. Tribble A.C. Cerne UE&C(SB-15990)

S.D. Floyd ASLB June 27, 1983 J.H. Herrin - R.J. Harrison W.P. Johnson J.A. MacDonald

, , SBN- 525 T.F. B7.1.2 c.F. Mcdonald T. Fuller D.N. Merrill P.L. Anderson D.E. Moody W.H. Fadden NRC Chrono A.E. Ladieu H.T. Tracy(2) G. Tsouderos United States Nuclear Regulatory Commission Projects-All Washington, D. C. 20555 D.A. Maidrand Projects-SLA Ropes & Gray (Dignan/Ritsher"tila-

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Attention: Mr. George W. Knighton, Chief A.M. Shepard Licensing Branch No. 3 J.W. Singleton Division of Licensing T.F. B7.1.2 G.S. Thomas

References:

(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) USNRC Letter, dated May 12, 1983, " Issuance of Supplement l No. I to the Safety Evaluation Report (Seabrook Station, Units 1 and 2)," G. W. Knighton to R. J. Harrison '

(c) PSNH Letter, dated April 14,1983, " Response to Generic

( Letter 82-33, Supplement No. I to NUREG-0737,"

J. DeVincentis to D. G. Eisenhut

Subject:

Emergency Classification System

Dear Sir:

In response to the open item delineated in Supplement No. I to the Safety Evaluation Report (Reference (b)), we have encloced a new Section 5.0 of the Seabrook Station Radiological Emergency Plan which provides a conceptual description of the Emergency Classification System.

Please note that the Emergency Action Level setpoints (Tables A.1-A.5) and Emergency Status Indicators (Tables A.1-A.5) color combinations are

, tentative (some also indicate that setpoints and color schemes will be provided later). Section 5.0 should be reviewed in light of its conceptual nature. The Westinghoase Owners Group, as of this writing, is continuing to l revise its Emergency Ra'sponse Guideli.nes (ERGS) in response to the NRC review. These ongoing changes to the ERGS are expected to effect the Emergency Action Level artpoints and Emergency Status Indicators color combinations.

Setpoints and color combinations will be provided subsequent to completion of the ERCS and Seabrook Station Emergency Operating Procedures (Reference (c) commits to December 1983 for completion of Emergency Operating Procedure s) .

1000 Elm St.. P.O. Box 330. Monchester. NH O3105 Telephone (603) 669-4000 . TWx 7102207595

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• t United States Nuclear Regulatory Commission June 27, 1983 Attention: Mr. George W. Knighton Page 2

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The enclosed Section 5.0 will be incorporated in OL Application Amendment 49.

Very truly yours, YANKEE A'lVMIC ELECTRIC COMPANY m ld 'sa John DeVincentis Project Manager ALL/pf I Enclosure cc: Atomic Safety and Licensing Board Service List

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Rep. Beverly Hollingworth Ms. Olive L. Tash

(- Coastal Chamber of e - ree Designated Representative of 209 Winnacunnet Road the Town of Brentwood Hampton, NH 03842 R.F.D. 1, Dalton Road Brentwood, NH 03833 William S. Jordan, III, Esquire Harmon & Weiss Edward F. Meany 1725 I Street, N.W. Designated Representative of suite 506 the Town of Rye Washington, DC 20006 155 Washington Road Rye, NH 03870 Roy P. Lessy, Jr., Esquire Office of the Executive Legal Director Calvin A. Canney U.S. Nuclear Regulatory Commission City Manager Washington, DC 20555 City Hall 126 Daniel Street Robert A. Backus, Esquire Portsmouth, NH 03801 116 Lowell Street P.O. Box 516 Dana Bisbee, Esquire Manchester, NH 03105 Assistant Attorney General Office of the Attorney General Philip Ahrens, Esquire 208 State House Annex Assistant Attorney General Concord, NH 03842 Department of the Attorney General Augusta, ME 04333 . Anne Verge, Chairperson Board of Selectmen

Mr. John B. Tanzer hwn Hall

( Designated Representative of South Hampton, NH 03842 the Town of Hampton 5 Morningside Drive Patrick J. McKeon Hampton, NH 03842 Selectmen's Office 10 Central Road Roberta C. Pevear Rye, NH 03870 Designated Representative of the Town of Hampton Falls Ruthanne G. Miller, Esquire Drinkwater Road Law Clerk to the Board Hampton Falls, NH 03844 Atomic Safety and Licensing Board U.S. Nuclear Regulatory Comunission Mrs. Sandra Gavutis Washington, D.C. 20555 Designated Representative of the Town of Kensington Dr. Maury Tye, President RFD 1 Sun Valley Association East Kingston. NH 03827 209 Summer Street l

Haverhill, MA 01830 5

Edward J. McDermott, Esquire l Sanders and McDermott Mr. Angie Machiros Professional Association Chairman of the Board of Selectmen 408 Lafayette Road Town of Newbury Hampton, NH 03842 Newbury, MA 01950 Jo Ann Shotwell, Esquire Assistant Attorney General f (' Environmental Protection Bureau Department of the Attorney General One Ashburton Place, 19th Floor Boston, MA 02108

( , SB1&2 Amendment 49 FSAR May 1983

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5.0 EMERGENCY CLASSIFICATION SYSTEM 5.1 Summary An Emergency Classification System has been defined which categorises a wide spectrim of component or system failures and other occurrences that could potentially reduce station safety margins. The incidents are categorized according to severity in the following four classes: Unusual Event, Alert, Site Area Emergency, and General Emergency.

These predetermined emergency classes are declared by Seabrook Station per-sonnel. They assist emergency response organizations in determining the assessment, corrective and protective actions to be taken onsite and offsite.

Emergency classifications are based upon events identified by certain measur-able and observable indications of station conditions. These indications of degrading station status are called Emergency Action Levels (EALs) and are listed in detail with their associated station conditions in Appendix A.

These EALs aid the operator in emergency recognition and assure the first step is completed in emergency response. It must be recognized that if conditions warrant such action, the classification of the event may change as the incident increases or reduces in severity.

5.2 Symptomatic Approach to Classific.ation

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A symptomatic approach has been developed to assist the operator in emergency recognition and classification. In order to concentrate the amount of plant process data provided to the operator to that which is necessary for event classification, use is made of the color coded status trees that are associated with the sympton based Emergency Operating Procedures. These symptomatic status tree analyses allow the operator to recognize accident severity and concentrate on the appropriate corrective actions. Symptomatic status trees which relate to Seabrook Station EALs are provided along with a description of the approach in Appendix A.

5.3 Emergency Classes 5.3.1 Unusual Event AN UNUSUAL EVENT INDICATES A POTENTIAL DEGRADATION OF STATION SAFETY MARGINS WHICH IS NOT LIKELY TO AFFECT PERSONNEL ON-SITE OR THE PUBLIC OFF-SITE OR RESULT IN RADIOACTIVE RELEASES REQUIRING OFF-SITE MONITORING.

Unusual Events are conditions which do not cause serious damage to the station and may not require a change in operational status. For a complete list of the Unusual Event conditions, refer to Appendix A, Table A.I.

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SB 1 & 2 Amendment 49

FSAR May 1983

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Station response and off-site notification associated with this event classification assure that sufficient emergency response personnel, both on and off site, are mobilized and respond to event conditions. Actual releases of radioactivity which substantially exceed Technical Specification limits may be involved, and thus radiation monitoring and dose projection may be an integral portion of the emergency response required. For a complete list of Alert conditions refer to Appendix A, Table A.2.

5.3.3 Site Area Emergency A SITE AREA EMERGENCY INDICATES AN EVENT WHICH INVOLVES LIKELY OR ACTUAL MAJOR FAILURES OF STATION FUNCTIONS NEEDED FOR THE PROTECTION OF THE PUBLIC.

The events included in this Site Area Emergency category represent a potential for off-site releases which could impact the public to the extent that pro-tective actions may be necessary. For a complete list of Site Area Emergency conditions, refer to Appendix A, Table A.3. ,

5.3.4 General Emergency A GENERAL EMERGENCY INVOLVES SUBSTANTIAL CORE DEGRADATION OR MELTING WITH PCTENTIAL FOR LOSS OF CONTAINMENT INTEGRITY.

For a complete list of General Emergency conditions, refer to Appendix A, Table A.4.

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SB 1 & 2 Amendment 49 FSAR May 1983

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APPENDIX A EMERGENCY CLASSIFICATION SYSTEM AND EMERGENCY ACTION LEVELS Tables A.1 through A.4. represent the station conditions and associated Emergency Action Levels (EALs) that are categorized in accordance with the four emergency classes. The emergency conditions include a wide spectrum of events that represent varying degrees of threat to station personnel onsite or the public offsite. As the tables show, full use is made of the various degrees of challenge to the five Critical Safety Functions (CSFs);

1) Suberiticality,

2) Core Cooling,

3) Heat Sink,

4) RCS Integrity, and

5) Contain:nent Integrity.

EALs relate levels of challenge to the CSFs and other numerous process para-meter indicators of emergency conditions such as system pressures, liquid levels, radiation intensity, and temperatures, to appropriate emergency classifications.*

• k Symptom based status trees simplify the initial emergency classification process by relating the sost critical safety parameter indicators directly to the EALs. The individual status trees for each of the five emergency con-ditions used in conjunction with Table A.5 assist the operator in emergency classification and also directs them to the appropriate Emergency Operating Procedures for mitigation of the incident. Symptomatic status trees indicating emergency conditions are available to the operator on the plant process com-puter and are displayed on SPDS. Hard copies of these status trees are also available.

The CSF status trees, Figures A.1 through A.5, are based on plant events which pose a threat to the safety status of the plant. Color coding is used to identify event priorities for the individual branches of the status trees as follows:

OCREEu -

The Criticai SafetF Function is satisfied - no operator action is called for.  !

hYELLOW- The Critical Safety Function is not fully satisfied -

operator action may eventually be needed.

hORANCE- The Critical Safety Function is under severe challenge -

prompt operator action is necessary, hRED -

The Critical Safety Function is in jeopardy - immediate I operator action is required.

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sa1&2 Ameadosot 49 FSAR May 1983

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l k Table A.5 provides a cross reference which sor'relates the CSFs by selor priorities and other emergency conditions by catesory with their appropriate essergency classes. For azample, if parameters of critical safety function Number 2., cora cooling, complete as orange branch of the appropriate fault tree then seeerding to Table A.5, a Site Area Energency elassifiestion is reached. However, if an orange branch of the core Cooling status tree is soupleted in conjunction with a 3 (Esat Sink) red or orange or a 5 (Contain-neat Integrity) red, then a General Emergency is reached as shown in,C61ona 3 CSF/ Combinations of Table A.S. Combinations of Csys and other emergency coaditions are correlated with emergency classes in coluem 4. Coltaan 5 liste emergency conditions which constitute emergency classifications independant of Critical Safety Functions.

• Numbers which have been provided are tentative and will be operationally verified. Evabers which have not been provided are to be calculated and verified, and will be provided at a later time.

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TABLs a UNUSUAL gVENT CtASSIFICATION Critical Safety Function / georgency Status amergency Condition seergency Action Level Indicatorse core Cooling o Core esit TCs less than 1200eF, RCS subcooling less then (later)oF, 2 yellow at least one RCP is running g RVLIS wide range is greater thang

• (Aster) 1 4 RCP (later) I 2 R'CP (later) Z 3 RCP (later) I I RCP l

l e No RCFs running, RCS subcooling less than (later)*F, core exit 2 yellow TCs is less than 700* d RVLIS narrow range greater than (later)!: ,

RCS Integrity o RCS Cold Leg temperature greater them (later)oF M less than 4 yellow (later)*F and RCS coolant system has exceeded 1000F per hour cooldown rateg o RCS pressure greater than cold overpressure limit, RCS temperature 4 yellow less than 305'F, and RCS temperature decrease less than 1000F per inourg Heat Sink o Pressure less than 1255 peig iniall SCa but greater than 1185 pois. 3 yellow o Pressure less than 1255 reig in all SCs with narrow range level 3 yellow not less than 34.51 an all SCs.

o Pressure less than 1185 pois in all SCs with narrow range level 3 yellow less than 201 in all SCs. ,

"se Containment e Contalesent radiation levels greater then (later): $ yellow f,s =-

Loss of Plant Process Computer o As indicated or observed 3 13. **

Loss of Of f site AC Power o As Indicated or observed; 9.

Radiological Releases o Releases exceeding Technical Specifications 6a.

Fire o Fire within the station protected area which requires outside fire- Ila, fighting essistanceg Control Room svacuation o With control remaining at remote safe shutdown panell 12a.

. Airplane crash, Train Derailment o By observation; 14.

or gaptosion Onsite Offsite Medical Assistance Required o seergency transport of the worker to local support hospitalg l$.

• for Contaminated and Injured Worker at Local Support Hospital Loss of Onsite AC Power Capability e As indicated or observed: 16.

Severe or Natural Phenomenon o Response spectrum seismic unit triggered; I7a.

I o 50 year flood or low water level by observation or receipt of 17d. *f g[

warning from offsite authoritiest R

, o Tornado observed onsiteg 17e. g ,.

o Marricane observed onsite (sustained winds of (later) for (later) 17f. **

period of tieeg o Security compromiseng 17h.

o Technical Specifications surpsesed cousing shutdown. 18.

• T . l. sedU 2s te T shl.* 4.5 amt FiRwr. e 4.1-4.4. ,

R r

• TABLE n.2 ,

ALERT CLASSIFICATION Critical Safety Function / geerseecy Status Emergency condition Emeraency Action Level Indicators

• Suberiticality o latermediate rense SUR is sero or positive in power range less then I orange SI when the reactor should be subcriticall RCS letegrity o RCS cold leg temperature le less than (later)OF, RCS pressure- 4 orange temperature point to right of lielt A (see Figure A.4) and RCS temperature decrease greater than 1000F per hourl ,

Core Cooling g Containment o When containment radiation to greater than (later) and, eithers 5 yellow.

Integrity No RCPe are running, core emit TCS are less than 7000 and 2 yellow RVLis narrow is greater than (later)!! or Core exit TCe less than 12000F, RCS oub70oling less than (later)*F 2 yellow

. at least one RCP is running, and RYLIS wide range le greater thans 1

(later)I 4 RCP (later)I 2 RCP (later)I 3 RCP (later)I I RCP .

e Containment lategrity and o Failure to isolate containment as indicated by Phase A and 7 and Failure to isolate ConUInnent Phase B isolation indication panels when containment radiation S yellow

• is greater than (later)I 5

Radiological Releases ,, o 10 time Technical specificationel 6b. 3 E e-Fire o Controlled fire which effects only one train of safety-related lla. u equipment with the potential for affecting the other traing Severe or Natural Phenomenon o garthquake Breater than containment foundation Dgg alara 17b.

Initiating a shutdown! .

o Tornado observed onsite which has degraded safety componente. 173 E

• I

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• To be used with Table A.S and Figures A.1-A.4. O$

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SITE AREA EMERCENCY CLASSIFICATI0lt critical Safety Function /

Emeraency Condition geergency Status Emergency Action Level indicators

• Subcriticality o Power range is greater than 52 power wnen reactor should I red be suberiticals Core Cooling o Core emit TCs less than 12000F but greater than 7000F, RCS

• 2 orange subcooling less than (later)*F, and no RCFs are running, and RYLIS marrow rante is greater than (later)23 o RCS subcooling less than (later)*F, no RCFs running, core 2 orange emit TCs less than 7000, and RVLIS narrow range less then (later)2s o Core exit TCe less than 12000F, and RCS subcooling less than 2 orange (later)*F, at least one RCP running, and RYLIS vide range less thang (later)I 4 RCP (later)I 2 RCP (later)I 3 RCP (later)I I RCP =

Meat Sink o Wide range level less than top of U-tubes in all SCs and total 3 red feedwater flow to SCs less then 670 spas o Wide range level less than top of U-tubes in all SCe sad total 3 orange feedwater flow to SCs greater than 470 spo and pressure greater

• than 1255 pois in all SCsg o Wide range level greater than top of U-tubes in at least one SC 3 orange and pressure greater then 1255 pois in all SCs3 g

• ==

RCS Integrity o RCS temperature decrease greater than 100 0 F in last 60 minutes d 4 red RCS Pressure-Temperature ratio esceede Limit A (see Figure A.4) ([

containment Integrity o containment pressure greater than 52 poigg 5 red o Containment pressure less than 52 peig,,but greater than 5 pois 5 orange a o Containment pressure less than 52 pais and containment sump not 5 orange less then (later);

Radiological Releases o Radiological releases exceed EPA PACS at Site Boundary 3 6c.

Loss of all AC Power o As indicated or observed; 3.

Loss of all DC Power o As indicated or observed: 10.

Fire o Uncontrolled fire which affects safety-related equipments llb.

Control Room Evacuation o Evacuation of control room without control at remote shutdown panelg 12b.

Severe or natural Phenomenon o Earthquake with potential impact on SSE. 17c.

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• To he ..se.1 with Table A.5 and Figures A.1-A.4.

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TABLk . 4 CENERAL EMERCENCY CLASSIFICATION Critical Safety Function / .

guertency Condition EmerRency Action Level Emergency Statue

_ ladicatorse Core Cooling o Core emit TCe higher than 12000F3 2 red o Core emit TCs less than I200*F but greater than 700*F, RCS embcooling less them (later)*F, no 2 red RCPs running and RVLIS marrow range less than (later)It Core Coating and Meat sink o Core emit TCs less than 12000F, RCS subcoollag less than (later)'F, no RCPs running, core emit 2 orange TCe greater than 700*F and RVLIS marrow range greater than (later)I combined with heat elsk wide 3 red range level less than top of U-tubes in all SCe and total feedwater flow to SCs less than 470 speg or Combined with wide range level greater than top of U-tubes is at least one SC and pressure not 3 orsage less than 1255 peig in all sca or ~

Wide range level not greater then top of U-tubes in any Sco and total feeduster flow to SCs 3 orange greater than 470 spa, and pressure not less then 1255 pois in all SCs o Core emit TCs less than 1200*F, and RCS subcooling less than (later)oF, no RCPs running, core 2 orange emit TCe less than 7000F, RVLIS narrow range less than (later)T combined with heat sink wide 3 red range level less than top of U-tubes in all SCe and total feedwater flow to SCs less thea 470 ,

SPeg or Combined with wide range level greater then top of U-tubes in at least one SC and pressure not 3 orange less than 1255 peig in all sces or Wide range level not greater than top of U-tubes in say SCa and total feedwater flow to SCe 3 orange greater than 470 spe, and pressure no less then 1255 peig in all Sces o Core exit TCe less than 1200*F, and RCS subcooling less than (later)OF, at least one RCP 2 orange running, RYLIS wide range less thans ,

(later)I 4 RCP (later)I 2 RCP n'

UI ==

(later)I 3 RCP (Ister)I 1 RCP 5e Combined with heat elak wide range level less than top of U-tubes in all SCe and total feed- 3 red **

water flow to SCs less than 470 speg or Combined with wide rease level greater than top of U-tubes in at least one SC and pressure 3 ersage not less then 1255 peig la all scos or-Wide range level not greater than top of U-tubee In any 3Ce and total feedwater flow to SCe 3 orange greater thea 470 spe, and pressure not less than 1255 peig ja all sca:

Core Cooling and containment o Contalement pressure greater than 52 pela combined with core emit TCe less than 1200*F, RCS Integrity 5 red and subcooling less than (later)*F, no RCPe running, core exit TCe greater than 7000F and RTLIS 2 orange marrow range greater than (later)I3 or ,

o With core emit TCe less than 1200*F, RCS embcooling less than (later)*F, no RCFs runeing, core 2 orange emit TCe less than 7000F, RVLIS narrow range less than (later)! 3 or o with core emit TCe less than 1200'F, RCS subcooling less thea (later)*F, at least one RC 2 orange running and RYLIS vide range less thans (later)I 4 RCP (later)I 2 RCP (later)I 2 RCP (later)I I RCP Containment latogrity and o Containment pressure greater than 52 peig combined with lose of all ac power as indicated or Loss of all AC Power 5 red observedi and 8.

o Containment pressure less than 52 peig but greater than 5 pelg combined with loss of all sc 5 orange power as indicated or observed; and 3. >

Containment Integrity and o ConteZneent pressure greater than 52 pais combined with lose of all a,c power as indicated or E Failure to lactate 5 red obserweig and 7. gR Cantainment ,, 3 8

Loss of All (C and DC Power o As indicated or observed. Ie 8 and 10. * * *

• To be used with Table A.5 and Figures A.1-A.4.

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SB 1 & 2 Amendment 49

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/

FSAR May 1983 TABLE A.5 EMERCENCY CLASS CROSS REFERENCE WITM CRITICAL SAFETT FUNCTIONS (CSFs)/EMERCENCY CONDI Emergency Class CSF Singles CSF/ Combinations Complications Miscellaneous -

Ceneral 2 red 2 orange /3 red 5 red /8 10+ 8 2 orange /3 orange 5 orange /8 2 orange /5 red 3 red /8 Site Area 5 red /7 i red 6c, 8, 10, 2 orange 11b, 12b, 17c 3 red

• 3 orange 4 red ' <15

,Y' ' ,r )S'"'

5 red 5 orange f

Alert 1 orange, 2 yellow /5 Yellow 5 yellow /7 hh* I I ' 8lb-4 crange IIK

\

Unusual Event 2 yellow 6a. 9, lle. 12a. '

3 yellow 13, 14, 15, 16, 4 yellow 17a, 17d. 17e.

5 yellow 17f. 17h. 18 critial Safety Functions

1. Suberiticality *

( 2. Core Cooling

3. Reat Sink

4. RCS Integrity *

5. Containment Integrity Hiscellaneous Emergency Conditions

6. Hi Releases a) Technical Specification b) 10 Techsical Specification c) Indications EPA PACS will be exceeded at site boundary

7. Failure to isolate containment

8. Loss of all ac power

9. Loss of offsite ac power

10. Loss of all de power l 11. Fires t

a) Controlled-offects only one train of safety-related equipment with l

' the potential for affecting the other train b) Uncontrolled-effects safety-related equipment c) Within plant protected area which requires outside fire-fighting assistance

12. Control room evacuation a) With control at remote shutdown panel b) Without control at remote shutdow6 panel

13. Loss of plant process computer 14 Observation of aircraft crash, train derailment or explosion onsite

15. Energency

16. Loss of onsite transport of contaminated and injured worker to local support hospital ac power capability

17. Severe or natural phenomenon a) Response spectrum seismic unit triggered b'

Earthquake greater than containment foundation DBE alare levels c) Earthquake with potential impact on SSE d)

( 50 year flood or low water level by observation or receipt of warning free offsite authorities e) Observation of tornado onsite .

f) Observation of hurricane onsite 7 g) (sustained winds of (later) for (later) period of time) h)

Tornado observed onsite which has degraded safety components Security compromises

18. Shutdown-Technical Specifications surpassed .

i RED

( _

ORANGE YELLOW POWER RANGE N

• LESS THAN -

5% y INTERMEDIATE N RANGE SUR MORE NEGATIVE THAN .2 DPM Y INTERMEDIATE N RANGE SUR -

ZERO OR  !

NEGATIVE Y -

( -

GREEN N

~

SOURCE RANGE -

ENERGlZED YELLOW SOURCE RANGE N

- SUR NEGATIVE -

OR ZERO y GREEN

( FIGURE A.1 STATUS TREE FOR CRITICAL SAFETY FUNCTION NUMBER 1 - SUSCRITICALITY

_ RED RED N

RVLIS NARROW RANGE GREATER -

THAN (2*)% y CORE EXIT N

--* TCs LESS -

THAN 1200* F y ORANGE CORE EXIT N TCS LESS -

THAN 700* y ORANGE AT LEAST N N.

RVLIS NARROW ONERCP - -

RANGE GREATER RUNNING y THAN (2*)% y YELLOW RCS N SUSCOOLING -

GREATER THAN ORANGE RVLIS WIDE RANGE N GREATER THAN (3* -

(4*)% 4 RCP

)% 3 RCP (5*)% 2 RCP Y (6,)% 1 RCP YELLOW GREEN

• SEE FOOTNOTES FOR CRITICAL SAFETY FUNCTION STATUS TREES

('

FIGURE A.2 I

STATUS TREE FOR CRITICAL SAFETY FUNCTION NUMBER 2- CORE COOLING

/P i RED

(

TOTAL FEEDWATER N FLOW TO -

SGs GREATER y

. THAN 470 GPM WIDE RANGE ORANGE LEVEL GREATER N i

^

THANTOPOF -

\

U-TUBES (1 *) N Y PRESSURE LESS IN AT LEAST ONESG l THAN 1255 PSIG -

lN ALL SGs y l

YELLOW PflESSURE LESS N THAN 1185 PSIG -

IN ALL SGs y

( ,

YELLOW

'~

1 NARROW RANGE N LEVEL LESS ._

THAN 84.5% IN ALL SGs Y _

r YELLOW NARROW RANGE N' ~ ~[ '

LEVELS GREATER

~

'THAN 20% IN

. j s, '- .,

, .. ALL SGs y w  ;,,.

m. _

• SEE FOOTNOTES FOR CRITICAL SAFETY ~' 4 .

% GREEN FUNCTION STATUS 1REES --

.s,

g ,.; -, -c .

4.,

k,.

] FIGURE A3 -

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4,:

s- STATUS TREE FOR CRITICAL 5 ACETY' FUNCTION '*

- ~,..i ,

NUMBER 3 - HEAT Sih K

. s a

gg, -

s. s

~ . . . .

s

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A.

12 8

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T1 T2 TEMPERATURE RCS PRESSURE TEMPERATURE N POINT TO -

RIGHT OF Y

LIMIT A ORANGE l

RCS COLD LEG N TEMPERATURE GREATER THAN (1 *)* F Y YELL'OW I RCS ^

k. TEMPERATURE N

• RCS COLD LEG N .

_ DECREASE LESS _,_

_ TEMPERATURE THAN 100* F IN GREATER THAN THE LAST 60 Y (2 *)* F Y MIN.

GREEN YELLOW RCS PRESSURE l N LESS THAN COLD -

TEM TURE Y IM T GREATER THAN 30S*F Y GREEN

(

• SEE FOOTNOTES FOR CRITICAL SAFETY FUNCTION STATUS TREES FIGURE A.4 STATUS TREE FOR CRITICAL SAFETY FUNCTION wieu msa 4 . n m sN Tsa nivv

(

RED W

CONTAINMENT N

% PRESSURE LESS -

THAN 52 PSIG y ORANGE CONTAINMENT N PRESSURE LESS -

THAN 5 PSIG y ORANGE

, CO'NTAINMENT N k SUMP LEVEL -

LESS THAN (1 *) Y YELLOW CONTAINMENT N RADIATION _,

LESS THAN (2*) Y GREEN

• SEE FOOTNOTE 4 FOR CRITICAL SAFETY FUNCTION STATUS TREES FIGURE A.5 STATUS TREE FOR CRITICAL SAFETY FUNCTION NUMBER 5 - CONTAINMENT m m "1 '

5 -

' SB1&2 Amendment 49

, FSAR - May 1983 k

FOOTNOTES FOR CRITICAL SAFETY FUNCTION STATUS TREES Figure A.2 Core Cooling (1) Enter sum of temperature and pressure measurement system errors translated into temperature using saturation tables.

(2) Enter plant specific value which is 3-1/2 feet above the bottom of active fuel in core with zero void fraction, plus uncertainties.

(3) Enter plant specific value corresponding to an average system void fraction of 50 percent with 4 RCPs running.

(4) Enter plant specific value corresponding to an average system void fraction of 50 percent with 3 RCPs running.

(5) Enter plant specific value corresponding to an average system void fraction of 50 percent with 2 RCPs running.

(6) Enter plant specific value corresponding to on average system void fraction of 50 percent with 1 RCP running.

Figure A.3 Heat Sink (1) Actual indicated level ~ corresponding to top of U-tubes is dependent on calibration of wide range channel performed prier to startup.

Figure A.4 RCS Integrity (1) Enter plant specific temperature corresponding to temperature T1 (refer to FR-P.1 background document).

(2) Enter plant specific temperature corresponding to temperature T2 (refer to FR-P.1 backgroend document).

Figure A.5 containment (1) Enter plant specific level corresponding to the combined volumes of:

RWST + Accumulators + RCS + 1/2 CST (to be calculated at a later date).

(2) Enter plant specific value corresponding to radiation level alare setpoint for post acciden't containment radiation monitor (to be set prior to initial plant operation).

O

\

a CERTIFICATE OF SERVICE I, R. K. Gad III, one of the attorneys for the Applicants hereih, hereby certify that on July 15, 1983, I made service of the within " Applicants' Direct Testimony No. 1" by mailing copies thereof, postage prepaid, to:

Helen Hoyt, Chairperson Diana P. Randall Atomic Safety and Licensing 70 Collins Street Board Panel Seabrook, NH 03874 U.S. Nuclear Regulatory .

Commission Washington, DC 20555 Dr. Emmeth A. Luebke William S. Jordan, III, Esquire Atomic Safety and Licensing Harmon & Weiss Board Panel 1725 I Street, N.W.

U.S. Nuclear Regulatory Suite 506 Commission Washington, DC 20006 Washington, DC 20555 Dr. Jerry Harbour G. Dana Bisbee, Esquire l Atomic Safety and Licensing Assistant Attorney General Board Panel Office of the Attorney General U.S. Nuclear Regulatory 208-State House Annex Commission Concord, NH 03301 Washington, DC 20555 Atomic Safety and Licensing Roy P. Lessy, Jr., Esquire Board Panel Office of the Executive Legal U.S. Nuclear Regulatory Director Commission U.S. Nuclear Regulatory Washington, DC 20555 Commission Washington, DC 20555 Atomic Safety and Licensing Robert A. Backus, Esquire Appeal Board Panel 116 Lowell Street U.S. Nuclear Regulatory P.O. Box 516 Commission Manchester, NH 03105 Washington, DC 20555 Philip Ahrens, Esquire Anne Verge, Chairperson Assistant Attorney General Board of Selectmen Department of the Attorney Town Hall General South Hampton, NH Augusta, ME 04333

a F D;vid R. L wic, E quiro Jo Ann Shotwall, Ecquirb Atomic Safoty cnd Licen2ing A3cicttnt Attorn;y G2noral Board Panel Environmental Protection Bureau U.S. Nuclear Regulatory Department of the Attorney General !

Commission One Ashburton Place, 19th Floor l Rm. E/W-439 Boston, MA 02108 l Washington, DC 20555

)

. Mr. John B. Tanger Ms. Olive L. Tash Designated Representative of Designated Representative of the Town of Hampton the Town of Brentwood 5 Morningside Drive R.F.D. 1, Dalton Road Hampton, NH 03842 Brentwood, NH 03833 Ms. Roberta C. Pevear Mr. Patrick J. McKeon Designated Representative of Selectmen's Office the Town of Hampton Falls 10 Central Road Drinkwater Road Rye, NH 03870 Hampton Falls, NH 03844 Mrs. Sandra Gavutis Mr. Calvin A. Canney Designated Representative of City Manager the Town of Kensington City Hall RFD 1 126 Daniel Street East Kingston, NH 03827 Portsmouth, NH 03801 Senator Gordon J. Humphrey Mr. Angie Machiros U.S. Senate Chairman of the Washington, D.C. 20510 Board of Selectmen (Attn: Tom Burack) Town of Newbury NeWbury, MA 01950 Senator Gordon J. Humphrey Mr. Richard E. Sullivan 1 Pillsbury Street Mayor Concord, NH 03301 City Hall (Attn: Herb Boynton) Newburyport, MA 01950 Mr. Donald E. Chick Mr. Maynard B. Pearson Town Manager 40 Monroe' Street Town of Exeter Amesbury, MA 01913 10 Front Street Exeter, NH 03833 Brian P. Cassidy, Esquire Regional Counsel Federal Emergency Management Agency - Region I 442 POCH Boston, Ma. 02109 r

F

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g 'l ___ s ' t R. K. Gad JOI